22.251 System Analysis of the Nuclear Fuel Cycle
Fall 2009
Homework Set #3
Problem 1. Consider three reactor types, a current large PWR, a proposed CANDU
fueled with slightly enriched U (SEU), and a small modular pebble-bed HTGR, having
the following characteristics:
PWR
PBMR
CANDU-ACE
Thermal Power, MWe
1150
114
881
Electric Power, MWth
3411
265
2798
FUEL ENRICHMENT, w/o U-235
4.5
8.0
1.20
DISCHARGE BURNUP, MWd/kg
50
80
19.75
FUEL Management
3-BATCH
CONTINUOUS
ON-LINE
REFUELING
CONTINUOUS
ON-LINE
REFUELING
(a) Compare the uranium and separative work utilization, in MWde/kg UNAT and
MWde/kg SWU, for an enrichment plant tails of 0.25 w/o.
(b) Determine which fuel cycle is superior to the others among the PWR, PBMR and
CANDU, and explain why the possible reasons behind this behavior.
Problem 2. This problem aims at examining the thermal hydraulic limits and their
implications for the design of advanced fuels of LWRs
It is proposed to use a 13 x 13 internally and externally cooled annular fuel assembly
design to replace the 17 x17 externally cooled fuel assembly in a typical Westinghouse
PWR reactor. The power density in the original core is 100 kW/liter. The new fuel rod is
assumed to operate with the same heat flux on both the external and internal surfaces. For
simplicity, you may assume that the new fuel/clad gap thickness and clad thickness are
equal to their values in the original solid fuel 17 by 17 assembly. Dimensions of the
original assembly are given in the Table below.
2.1.a) If the external annular fuel rod diameter is 1.54 cm and the internal diameter of the
annular fuel rod is 0.8cm, what is the enrichment of the fuel that will conserve the
amount of U235 in the reload fuel if the current reactor uses 4.5% enrichment. This will
allow the innovative fuel to operate at the same power density for roughly the same fuel
cycle length. You may assume that the clad and gap thicknesses on the inside or outside
of the annular fuel are the same and equal to those of the solid fuel.
2.1.b) Explain why the actual enrichment needed to operate at the same core power
density for the same fuel cycle length will be higher than what was estimated in part
2.1.a.
2.2.a) If the heat flux is assumed constant on the outer and inner surfaces of the annular
fuel rod, what is the heat flux at the cladding surface in the new fuel assembly when it
operates with the same power density as the old fuel assembly?
2.2.b) At what conditions would the heat fluxes at the inner and outer surfaces become
equal? Evaluate, qualitatively, the validity of this approach.
2.3.a) Can we increase the power density in the new core if we want to maintain the same
Minimum Critical Heat Flux Ratio (MCHFR) as for the old fuel? The designer wishes to
keep the power to coolant flow ratio in the core at its value in the old design. You may
assume the Critical Heat Flux in subcooled and low quality flow boiling in water to be
proportional to G0.4/De0.6, where G is the coolant mass flow rate per unit area and De is
the equivalent hydraulic diameter. For this solution, you may deal with the whole
assembly as a single hydraulic channel and the heat flux is identical on the outer and
inner surfaces of the annular fuel rod.
2.3.b) Is assuming the whole assembly as a single hydraulic channel valid? Evaluate,
qualitatively, the validity of this assumption.
2.4) Assuming that the fuel cycle length remained the same, what are the consequences of
operating at higher power density in the new core for each of the following: (i) initial
enrichment needed, (ii) the fast neutron fluence of the cladding at discharge (iii) The
decay heat level in the discharged assembly?
Table 1: Assembly parameters for a Westinghouse 17X17 assembly
Assembly pitch (cm)
21.5
Lattice pin pitch (cm)
1.26
Fuel pellet radius (cm)
0.4096
Gap thickness (cm)
0.0082
Clad thickness (cm)
0.0572
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22.251 Systems Analysis of the Nuclear Fuel Cycle
Fall 2009
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